This invention relates to a fuel assembly for a nuclear reactor and more particularly to a fuel assembly for a light water reactor such as a boiling water reactor capable of ensuring a long operation cycle and maintaining a high shut-down margin.
A fuel assembly for a boiling water reactor (BWR) is constructed by a square channel box in which a number of fuel rods each comprising a metallic clad in which nuclear fuel material is packed are systematically arranged. The reactor core of the BWR includes a plurality of cells each comprising a cruciform control blade and four fuel assemblies surrounding the control blade and these cells are arranged in a systematic manner in the core. That is, each fuel assembly and control blade have axes perpendicular and parallel to each other, and a coolant operated as a moderator flows from the lower portion towards the upper portion of the reactor core.
Concerning the BWR, steam voids are not formed in a portion near the lower end of the core effective portion, i.e., the lower end of a heat generating portion at which an exothermic reaction is performed, but a large number of voids are generated at the area above the central portion of the reactor core, and the generated voids move up towards the upper portion of the reactor core. Accordingly, the void fraction in the BWR becomes high towards the upper portion of the reactor core, and as a result, the moderation characteristics to neutrons are lowered, whereby the output power is also lowered. In order to obviate these defects, in a conventional technique, it has been a practice to increase the enrichment of the fissile nuclide to be contained in the fuel at a portion of high void fraction, or it has been a practice to mix a burnable poison with the fuel element to suppress the increasing of the power output at a portion of low void fraction.
For the reasons described above, in the BWR, the burn-up at the upper portion of the core is liable to be delayed, and hence the concentration of U-235 becomes relatively higher than that of the other portion of the core. In addition, since a fissile nuclide such as Pu-239 is generated by the voids, it is difficult to maintain the shut-down margin of the reactor core at the upper portion thereof. Moreover, recently, many endeavors have been made for elongating the reactor operation cycle of the reactor and improving the degree of burn-up of the fuel in order to satisfy the economical requirement. In these cases, however, the enrichment of the fuel is necessarily increased, so that the maintenance of the shut-down margin of the reactor is made further difficult.
The fuel assembly which has been conventionally used and a fuel assembly which is expected to be used in the near future for the boiling water reactor (BWR) will be described hereunder by way of typical examples with reference to the drawings.
FIG. 31A is a perspective view of a fuel assembly of conventional type and FIG. 31B is a schematic vertical sectional view of a fuel rod consisting of the fuel assembly.
Referring to FIG. 31A, the fuel assembly comprises water rods, not shown, and fuel rods 2 secured by an upper tie plate 4, a spacer 5 and a lower tie plate 6, and a channel box 1 surrounding the outer periphery of the thus secured water rods and fuel rods 2. Each of the fuel rods 2, as shown in FIG. 31B, comprises a clad or sheath 7, a plurality of fuel pellets 8 arranged in the clad 7, a spring 9 located in a gas plenum disposed above the pellets 8 in the clad 7, an upper plug 10 for closing the upper opening of the clad 7, and a lower plug 11 for closing the lower opening of the clad 7.
FIG. 32 is a cross sectional view of the conventional fuel assembly shown in FIG. 31, in which sixty-two fuel rods 2 and two water rods 3 are arranged in the channel box 1 to constitute the fuel assembly. The water rods 3 serve to suppress the shortage of the water acting as the moderator in the interior of the fuel assembly, but the water rods 3 are axially uniformly arranged, so that there may arise such problems as excessive water condition at the lower portion of the reactor core or water shortage condition at the upper portion thereof.
FIG. 33 also is a cross sectional view of a fuel assembly which has been developed for improving the characteristics of the fuel assembly shown in FIG. 32. The fuel assembly shown in FIG. 33 includes one water rod 12 having a diameter larger than that of the water rod 3 to pass non-boiling water therethrough. However, even in this example, there arises a problem of the excessive water condition at the lower portion of the reactor core and the water shortage condition at the upper portion thereof as described with respect to the former example shown in FIG. 32.
FIG. 34 is a cross sectional view of a further example of a conventional fuel assembly developed for improving the fuel assembly of the type shown in FIG. 32. The fuel assembly of FIG. 34 comprises four square channel boxes 13 each containing sixteen fuel rods 2 which are arranged to constitute a water area of a boiling moderator material, and a cross-shaped space 14 defined by the respective channel boxes 13 constitutes a water area of a non-boiling moderator material to thereby seek a uniform distribution of the output power in the horizontal direction. With the fuel assembly of this character, however, there also arise problems of the excessive water condition at the lower portion of the reactor core and the water shortage condition at the upper portion thereof.
FIG. 35 is a cross section of a still further example of the conventional fuel assembly of a type which is an improvement over that shown in FIG. 34. The fuel assembly of FIG. 35 is constructed by nine sub-bundles 15 each comprising nine fuel rods 2, and relatively wide gaps 16 are defined between the respective sub-bundles 15. With the fuel assembly of this example, the problems of the excessive water condition and the water shortage condition at the lower and upper portions of the reactor core have not been solved.
As described hereinabove, concerning the BWR, steam voids are formed in the area of the heat generating portion of the BWR and the voids move up towards the upper portion of the reactor core. Accordingly, the void fraction in the BWR becomes high towards the upper portion of the reactor core and the average density of the water is lowered. The moderation characteristics relative to neutrons are lowered towards the upper portion of the reactor core and hence the fission rate is also lowered. In other words, the burning progresses at the lower portion of the reactor core and the burning is delayed at the upper portion thereof. In order to obviate this phenomenon for the purpose of suppressing the lowering of the output power at the upper portion of the reactor core, it has been proposed to design the reactor core to increase the enrichment of the fissile nuclide to be contained in the fuel disposed at the upper portion of the reactor core.
However, this design leads to an increase in the reactivity at the upper portion of the reactor core and hence reduces the subcriticality during the shut-down period of the BWR, and is therefore not desirable.
In the meantime, in order to improve the economical requirement by elongating the operation cycle or increasing the take-out burn-up degree, it is most effective to increase the enrichment of the fuel. However, in general, increasing of the fuel enrichment further reduces the subcriticality at the reactor shut-down period. Accordingly, when the fuel enrichment has been largely increased, the reactor may not be safely shut down during the low temperature operation period in certain cases. This point makes it difficult for the conventional fuel to elongate the operation cycle or increase the take-out burn-up degree. In addition, with the conventional fuel assembly, a ratio (H/U) of the number of hydrogen atoms H with respect to the number of uranium atoms U is distributed to be greater at the upper portion of the fuel assembly and smaller at the lower portion thereof due to the distribution of the voids during the power operation period. Accordingly, the average infinite multiplication factor in the reactor core is not made sufficiently high during the reactor power operation period, which is also a significant problem of the nuclear reactor.